In line with Polish national activities and research programs investigating non-electrical-reactor use, the national GOSPOSTRATEG-HTR project was launched, aiming at the development of a novel pre-conceptual design of a High-Temperature Gas-cooled Reactor (HTGR). The 40 MW_th research reactor would serve as a technology demonstrator for future industrial purposes. In the paper, the proposal of an established thermal-hydraulic and neutronic core design is presented as a result of the National Centre for Nuclear Research team studies, in the scope of the project, including the areas of fluid mechanics, heat exchange and reactor neutronic core design support analyses. The undertaken analyses were confirmed by the series of code investigations involving integral thermal-hydraulic (MELCOR (Sandia National Laboratories, USA), CATHARE (CEA, France)), neutronic (Serpent (VTT, Finland), MCB (AGH University’s Department of Nuclear Energy, Poland)), Computational Fluid Dynamics (ANSYS Fluent (ANSYS, USA)) and others. The calculations performed within the preliminary safety analysis on the pre-concept showed its compliance with international safety standards for the normal operation and Design Basis Accident sequences.

The work presented in this paper was performed within the Euratom Horizon 2020 GEMINI Plus project. Behavior of the HTGR reactor under severe accident conditions was investigated and the maximum fuel temperature was observed. Due to application of the TRISO particles and SiC layers in the fuel element, no damage of the fuel is expected up to 1600°C. Under the cooperation in the project between Nuclear Research Group (NRG) and National Centre for Nuclear Research (NCBJ) a code-to-code calculations were carried out between the SPECTRA and MELCOR codes. SPECTRA code, developed by the NRG is a thermal hydraulic analysis code and MELCOR 2.1.6342 used by NCBJ developed by SANDIA National Laboratory is fast running severe accident code. Both codes have already HTGR specific models build in. The following accident was analyzed and will be presented: Depressurized Loss of Forced Circulation (DLOFC) with 65 mm break at the top of reactor vessel. The scenario was calculated applying following sets of assumptions: best estimate and conservative. Plant behavior was analyzed including primary and secondary side of the reactor. As the results of applying conservative assumptions, it was found that fuel temperature excides the acceptable limit of 1620°C. Therefore, changes in the core design were proposed by project participants. Analyses of the new core showed acceptable temperatures. In the paper the results of code-to-code comparison are presented. Both codes have shown a good agreement of presented following characteristics on maximum fuel temperature, relative power and Reactor Cavity Cooling System power, primary pressure and break flow.